Development of a New Aqueous Process for Nuclear Fuel Reprocessing: Hot Tests on the Recovery of U and Pu from a Nitric Acid Solution of Spent LWR Fuel

Abstract

We have studied a new aqueous reprocessing system that consists of anion exchange as the main separation method, electrolytic reduction for reducing U(VI) to U(IV), and extraction chromatography for minor actinide partitioning. In this work, hot tests were carried out on the main flow sheet (U and Pu recovery) using a nitric acid solution of a spent commercial boiling water reactor fuel with burnup of 55 000 MWd/t HM. First, a separation experiment was conducted using a column packed with AR-01 anion exchanger, and the separation behavior of about 20 elements was examined. Then electrolytic reduction was performed for the U(VI) eluate from the first column using a flow-type electrolysis cell. Subsequently, the reduced U solution was applied to the second AR-01 column to separate the U(IV) from contaminated fission products. Most amounts of Pu(IV)-Np(IV), were successfully separated and recovered in the first column. In the electrolysis, U(VI), Np(V,VI), and a trace amount of Pu(VI) were reduced to U(IV), Np(IV), and Pu(IV), respectively. In the second column, the U(IV) with small amounts of Np(IV) and Pu(IV) was completely separated from the fission products. These results demonstrated that the proposed U and Pu recovery process is essentially feasible, though more effective elution methods for Pd and Tc need to be investigated further.